openmsr / msr-archiveLinks
curated documents pertaining to early molten salt reactor research.
☆25Updated last year
Alternatives and similar repositories for msr-archive
Users that are interested in msr-archive are comparing it to the libraries listed below
Sorting:
- List of open source projects related to OpenMC☆30Updated 6 months ago
- ☆33Updated 5 months ago
- Automated V&V for fusion relevant CAD-based benchmarks☆23Updated 2 weeks ago
- Python ENDF Parser☆39Updated 2 months ago
- API for parsing and post-processing many MC simulations related files☆17Updated last week
- Accelerated Discretized Geometry☆22Updated 2 weeks ago
- JADE, a novel nuclear data libraries V&V tool☆34Updated last week
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆41Updated 4 years ago
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆18Updated 2 weeks ago
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆61Updated 2 months ago
- A two-cycle full-core PWR depletion benchmark based on a commercial nuclear power plant☆40Updated this week
- Nuclear data processing with legacy NJOY☆122Updated last week
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆48Updated last week
- Stochastic Calculator Of Neutron transport Equation☆53Updated last week
- Repository for running SINBAD benchmarks with OpenMC and UQ☆10Updated 3 years ago
- NJOY for the 21st Century☆84Updated last year
- Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections☆11Updated 3 years ago
- Tools to translate between different CSG geometry types☆40Updated 6 months ago
- Native plotting GUI for model design and verification☆64Updated last month
- Tool for converting MCNP input files to OpenMC classes/XML☆52Updated 2 weeks ago
- Convert OpenMC Geometry Components to CAD☆20Updated last year
- Catalogs a plethora of reactor inputs for OpenMC☆35Updated 9 months ago
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆14Updated 7 months ago
- For Updating Data and Generating Evaluations (FUDGE): LLNL code for managing nuclear data☆32Updated 2 months ago
- Open-source Sn software☆52Updated last week
- Workflow and Template Toolkit for Simulation (WATTS)☆33Updated 6 months ago
- Three-Dimensional Discrete Ordinates Neutron, Photon, Electron and Positron Transport Code☆11Updated 3 years ago
- code to target the conversion from a step-file to a h5m-geometry for neutronics☆35Updated 5 months ago
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆43Updated last year
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆83Updated 3 weeks ago